Active Control Coils Research Articles

Plasma control for the step prototype power plant

In 2019 the UK launched the Spherical Tokamak for Energy Production (STEP) programme to design and build a prototype electricity producing nuclear fusion power plant, aiming to start operation around 2040. The plant should lay the foundation for the development of commercial nuclear fusion power plants. The design is based on the spherical tokamak principle, which opens a route to high pressure, steady state, operation. While facilitating steady state operation, the spherical design introduces some specific plasma control challenges: (i) All plasma current during the burn phase should to be generated through non-inductive means, dominated by bootstrap current. This leads to operation at high normalised plasma pressure βN with high plasma elongation, which in turn imposes effective active stabilisation of the vertical plasma position. (ii) The tight aspect ratio means very limited space for a central solenoid, imposing that even the current ramp up must be non-inductively generated. (iii) The compact design leads to extreme heat loads on plasma facing components. A double null design has been chosen to spread this load, putting strict demands on the control of the unstable vertical plasma position. (iv) The heat pulses associated with unmitigated ELMs are unlikely to be acceptable imposing ELM free operation or active ELM control. (v) To reduce and spread heat loads, core and divertor radiation and momentum loss has to be controlled, aiming to operate with simultaneously detached upper and lower divertors. (vi) High pressure operation is likely to require active resistive wall mode (RWM) stabilisation. (vii) The conductivity distribution in structures near the plasma must be carefully selected to reduce the growth rates for the vertical instability and the RWM without damping the penetration of the of magnetic fields from active control coils too much. This article describes the initial work carried out to develop a STEP plasma control system.

Error field feedback control system in the Keda Torus eXperiment and open loop control experiment

A feedback control system has been designed for the Keda Torus eXperiment device to achieve the following goals: (a) suppression of the error field at the poloidal gap, (b) three-dimensional plasma stability control, and (c) improvement of plasma discharge quality. The system consists of a boundary electromagnetic probe array, a field-programmable gate array, a linear power amplifier, and an active control coil array. The system adopts a compound control method that includes both active and passive control methods. To control the plasma, an active control coil array is used to generate a localized radial magnetic field as needed at two poloidal gaps of the device. In the open-loop control experiment, the active control coils near the poloidal gaps reduced the original error field and thus increased the plasma discharge current amplitude by 50% and effectively extended the plasma discharge duration.

Studies of the plasma vertical instability and its stabilized concepts in JA and EU broader approach, DEMO design activity

Vertical instability of an elongated plasma and its stabilized concepts by in-vessel components and vacuum vessel (VV) design have been studied intensely in JA and EU Broader Approach, DEMO Design Activity. The vertical stabilization of the plasma represents one of the key issues for EU and JA DEMO, due to the large distance of the active control coils for the presence of thick breeding blanket system. A feasible DEMO reactor that maintains plasma vertical stability was proposed from an engineering viewpoint. The vertical stability performances are acceptable, without considering a passive stabilizer, if a maximum elongation of κ95 = 1.6 is chosen. For the higher-elongated plasmas (κ95 > 1.70), additional inboard passive stabilizer is effective.

Overview of Keda Torus eXperiment initial results

The Keda Torus eXperiment (KTX) is a new reversed field pinch (RFP) device at the University of Science and Technology of China. The construction and assembly of KTX, including the vacuum chamber, conducting shell, magnetic field windings, power supply system, active control coils, vacuum pump and data acquisition system, have been completed on August 1, 2015. Immediately following that, the first plasma was obtained on August 15, 2015. Intensive conditioning of the machine is underway to ramp up the plasma current toward its full operation. An active feedback mode control system has been built and has been implemented to control the error field around the vertical gaps of the conducting shell. The pulsed power supply systems of ohmic heating field and toroidal field (TF), using thyristor and energy storage capacitors, have been tested and commissioned. The TF power supply has flexibility in being able to operate with a reversed TF configuration and stable TF configuration. The fundamental diagnostic tools are developed for early KTX operation. Currently, the plasma current is up to 205 kA and the maximum discharge length is 21 ms, approaching to the conducting shell penetration time. Furthermore, typical RFP discharges are being produced with RFP state lasting 2 ms. These initial operation results for KTX are described in detail.

A constrained control strategy for the shape control in thermonuclear fusion tokamaks

The paper deals with the application of the so-called Reference (or Command) Governor constrained control strategy to the shape control of plasmas in thermonuclear fusion reactors with the main scope of optimizing tokamak operations also in conditions very close to the operating envelope limits. A primal inner loop controlling the plasma–wall distance is first designed; the Reference Governor device is then tuned to modify, whenever necessary, the reference signals to the inner loop, on the basis of constraints due to voltage saturations on the power supply converters, limitations of currents in the active control coils, minimum clearance between the plasma surface and the vacuum chamber wall, maximum induced magnetic fields and forces on coils. As usual in model predictive paradigms, the reference signal modification is accomplished through an on-line optimization procedure which embodies plasma model forecasts computed along a finite time virtual receding horizon. The ITER (International Thermonuclear Experimental Reactor) tokamak is assumed as the case study. Numerical simulations are carried out on a finite elements nonlinear model taking into account induced currents in the passive structures. The proposed application shows how almost a hundred constraints can be managed on-line by the Reference Governor.

RWM control studies on RFX-mod with a limited set of active coils

In this paper, the results of resistive wall mode (RWM) control experiments using several sets of active coils are presented.The control effectiveness as a function of coil number and covered torus surface is studied in the RFX-mod device, thanks to the powerful and flexible MHD active control system. Active control is tried using similar coil number and size both in the reversed field pinch (RFP) and in the tokamak configurations, giving a good experimental set to exploit the differences between the two cases.The experimental results are compared with predictions by a new integrated simulator for closed loop RWM growth. The numerical tool couples, in a self-consistent way, a full 3D finite element description of the machine boundaries, a 2D toroidal model of RFP plasma stability and a model of the RFX-mod control system, producing an overall dynamic model cast in the state variable space. In this way a full dynamic flight simulator of RWM control experiments is implemented, where coil coverage and experimental proportional–integral–derivative gains are explored in simulations.Both in the performed experiments and in simulations wall mode active control is proven to be possible in both the configurations with a very reduced set of active control coils.

Overview on the power supply systems for plasma instabilities control

The paper presents an overview on the power supply (PS) systems for plasma instabilities control in fusion experiments, based on active control coils. First, the MHD instabilities and the approach to their control in Tokamaks and Reversed Field Pinches (RFPs) are described. Then, the features of MHD modes controls presently used in fusion experiments are reviewed. For the control systems based on active coils fed by fast power supplies, the typical requirements in terms of power, dynamics, accuracy and delay are summarized and discussed. Then, a survey on the technology available to design these types of PSs is given, together with the most suitable circuit topologies and guidelines for the design, on the basis of solutions adopted in existing experiments.

Suppression of rotating external kink instabilities using optimized mode control feedback

Rotating external kink instabilities have been suppressed as well as excited in a tokamak using active magnetic coils that directly couple to the plasma through gaps in passive stabilizing conducting shells that surround the plasma. The kink instability has a complex growth rate, approximately (3+i2π5)×103s−1, and is near the ideal wall stability limit when discharges are prepared with a rapid plasma current ramp and adjusted to have an edge safety factor near 3. The active control coils are driven by a digital mode control feedback system that uses multiple field-programmable gate arrays to analyze signals from 20 poloidal field sensors and achieve high-speed feedback control. The feedback coil geometry used was designed to optimize feedback effectiveness. Signal processing is of critical importance to optimize phase transfer functions for control of rotating modes.

The Paralleling of High Power High Frequency Amplifier Based on Synchronous and Asynchronous Control

The vertical position of plasma in the HT-7U Tokamak is inherently unstable. In order to realize active stabilization, the response rate of the high-power high-frequency amplifier feeding the active control coils must be fast enough. This paper analyzes the paralleling scheme of the power amplifier through two kinds of control mode. One is the synchronous control; the other is the asynchronous control. Via the comparison of the two kinds of control mode, both of their characteristics are given in the text. At last, the analyzed result is verified by a small power experiment.

Control of the vertical instability in advanced tokamak

The control of vertical instability is studied for DN divertor plasmas in a tokamak with both passive and active control coils. The characteristic parameter of the vertical instability and a second damp stable operating region of the feedback gain are found. The favorable effects of higher triangularity and hollow profile on the critical elongation are demonstrated.

FEEDBACK STABILIZATION OF AXISYMMETRIC MODES IN NON-CIRCULAR TOKAMAKS

The feedback stability of axisymmetric MHD mode in tokamaks with unclosed shell has been studied analytically through solving directly the boundary problem of the perturbed magnetic surface in the vacuum region. For plasma torus with elliptical cross-section, a feedback system with conducting shell and active control coils has been suggested. The result shows that two unclosed shells arranged above and below the plasma have a very strong stabilizing effect. The re istive shell can suppress the modes with faster growth rate and the active coils can be used to stabilize the slower modes.